ML20211N493
| ML20211N493 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 12/11/1986 |
| From: | Ainger K COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2504K, NUDOCS 8612180235 | |
| Download: ML20211N493 (42) | |
Text
[
t
_ - One First Nabonal Plaza, Chca00, IEnois
\\
Address Reply to: Post Omco Box 767 Chca00, IEnois 60000 0767 December 11, 1986 Mr. Harold R. Denton, Director office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC. 20555
Subject:
Byron Station Units 1 and 2 Spent Fuel Pool Expansion NRC Docket Hos. 50-454 and 50-455
Reference:
November 25, 1986 letter from L.N. Olshan to D.L. Farrar
Dear Mr. Denton:
The referenced letter requested additional information regarding our proposed license amendment of September 3, 1986 to rerack the spent fuel pool at Byron Station. The responses to the request for additional information are enclosed with this letter.
Please direct any questions regarding this matter to this office.
one signed original and fifteen copies of this letter and enclosure are provided for NRC review.
Very truly yours, K. A. Ainger Nuclear Licensing Administrator 1m Enclosure cc: Byron Resident Inspector i
h hh 2504K 21 4
i I
R7eponst to NRC Quentions Item 1:
For the spent fuel pool heat exchaagers provide the following:
a.
heat transfer conductivity in BTU /Hr.
FTZ/OF b.
tube surface area in square feet
Response
a.
The heat transfer conductivity in BTU /HR FT4/0F Clean Condition:
501 Fouled Condition: 326 b.
The tube surface area is 6463 sq. ft.
1 u
Item 2:
Section 5.1 of the Licensing Report (Attachment B to the September 3, 1986 letter) indicated that Branch Technical Position (BTP) APCSB 9-2 was used, but Reference 1 to Section 5 indicated BTP ASB 9-2, Revision 2, July 1981 was used.
Correct this discrepancy.
Also confirm that the guidance in Standard Review Plan Section 9.1.3.III, regarding uncertainty factors, has been used.
BTP ASB 9-2, Rev. 2, July 1981 was used as cited in the reference.
The text in section 5.1 as in error on this point.
The use of uncertainty factors was in accordance with Standard Review Plan Section 9.1.3.III.
CECO RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION SPENT FUEL POOL RERACKING Item 3:
With respect to the spent fuel pool structure, the Licensee
-provided virtually no documentation attesting to the adequacy of
+
the analytical procedures, the load combination criteria, or the selection of allowable loads and stresses, other than to reference the original spent fuel pool analyses included in the FSAP..
Accordingly, the Licensee is requested to provide the following for review:
Provide sketches and/or drawings of any changes to the spent-A.
fuel pool structure not considered in the FSAR analysis.
Response 3A:
The installation of the high density fuel storage racks does 4
not necessitate any change to the spent fuel pool structure.
I The Byron high density fuel racks are free standing and apply loads directly to the bottom pool slab.
The bottom pool slab in a part of the building foundation basemat which j
is not highly stressed from the pool loads.
]
B.
Provide a description of the mathematical model of the pool structure, including the finite-element model if used, and the method of analysis.
Describe the assumptions employed and the limitations of the model.
Response 3B:
l The pool was analyzed by the finite element method of analysis using the SLSAP-IV computer program.
The pool slab and walls are modelled utilizing a mesh of quadrilateral i
plate /shell elements.
Spring boundary elements are used to model the supporting foundation material (Figure 1).
The assumptions made in the analysis are a)
Concentrated loads are distributed as nodal point loads.
b)
The stiffness of the foundation boundary elements is conservatively based on the Braidwood foundation properties.
The foundation at Braidwood is softer than Byron which results in higher stresses in the structure.
2 i
_,___-,-m.-
WEST WALL i
9 f
i
_ E L. 4 ol'- o '
l I
_ EL. Sec,8-o
E L A m '- 08 NORTH WALL EAST le Ih6) ip
/2 /
(g,, g, WALL p._____
/
,t_
I/
t'6 DEMOTES CRITICAL LOCATION
/
5 o'
/
W
)b BASE MAT s ! -A/
/
__ 4
/
/
r$-
3 EL.aeol o"
/
~
+
l
/
1,0
\\"
/
k E L.382'- o"
/
FIGURE I
?
l c)
Since the pool structure is symmetrical about column line 18,-half of the pool is modelled.
The nodal restraints along the line of symmetry are adjusted i
i.
depending on the loading condition to reflect symmetry or antisymmetry.
d.
Horizontal displacements are restrained for the east and west walls in the areas where the fuel handling building slab frames into-the walls.
C.
Provide ample description of the loadings used, and justification for the load combinations used.
+
Response 3C:
The following loading conditions are considered and used in the analysis:
a.
Thermal expansion of the liner b.
Horizontal hydrostatic pressure c..
Thermal gradient d.
Axial. expansion e.
Dead loads including fuel, fuel racks, self weight of i
structure, and vertical hydrostatic pressure f.
Dead loads from the cask pit side of the pool g.
Accidential cask drop h.
Hydrodynamic forces and seismic excitation of dead load for OBE and SSE The loading conditions above are combined into 15 load i
combinations per the B/B FSAR (Table 3.8-10).
The following controlling combinations are evaluated.
a.
Normal b.
Severe environmental c.
Abnormal (cask drop) 1, d.
Extreme environmental l
D.
Document the source of the acceptance criteria and metho'd of determining the ellowable loads and stresses in various parts of the structure.
i Response 3D:
\\
I The allowable stress acceptance criteria is found in the B/B
{
FSAR.
The computer program TEMCO is used to check the' reinforced concrete sections for the factored design moments and forces.
The resulting stresses are evaluated against s
l the following acceptance criteria:
il i
Maximum allowable steel stress = 54.0 ksi Maximum allowable concrete stress = 3,000 psi 1
i r, _._,
~
E.
Describe how the dynamic interaction between the pool structure and the rack modules was considered, including the key assumptions used in assessing the interaction effects and the value of any dynamic amplification factors.
Also include all assumptions made regarding the summation and phasing consideration of all rack module dynamic loads.
Response 3E:
Dynamic interaction between the pool structure and rack modules is not significant, and the rack analysis is performed assuming the pool structure is rigid.
The Byron high density fuel racks are free' standing racks which may slide and uplift under earthquake.
A nonlinear seismic analysis is performed on the racks to determine rack response.
The input to this analysis are the responses calculated from the overall building seismic model.
The Byron /Braidwood building seismic models are described in FSAR 3.7.2.3 and the fuel pool is included in the auxiliary-fuel handling-turbine building model.
The magnitude of the pool mass is small enough in comparison to participating building mass so that a change to this mass is not significant to the overall auxiliary-fuel-turbine building seismic response.
Therefore, overall dynamic _ interaction between.the pool constituents and the building is not significant.
i Evaluation of local interaction requires comparison of the local stiffness of the supporting structure and the rack stiffness.
Because the pool slab is a foundation slab, it is rigid in comparison to the rack fuel system and need not be included in the rack analysis.
Furthermore, since the racks are free standing, there is no significant interaction with the pool walls.
The results of the rack analysis are used to design the pool structures.
Dynamic amplification of the rack and fuel mass is accounted for in the dynamic analysis of the rack system.
The maximum rack vertical loads and horizontal loads are conservatively assumed to be in phase for the evaluation of the pool structure.
F.
Provide analysis of the adequacy of the pool floor and liner under rack sliding and impact loads.
1 l
,-,-r,,,r--w
_,nn-m,,-v,
,e----n,,-em
,8m-w,vv-me.,_--~~m---~~,~~~.,-+,-
~~
Response 3F:
The rack sliding and impact loads are included in the maximum support loads supplied by the rack vendor.
A local check of the pool floor and stainless steel liner is made for maximum rack foot impact loads by confirming the adequacy of these elements for the resulting bearing stresses.
Vendor supplied rack loads are also accounted for in overall behavior of the pool structure through use of the finite element analysis.
G.
Identify the critical regions of the pool structure.
List the loads or stresses as appropriate.
Carpare the loads and/or stresses to allowable values and indicate the source of the allowables in accordance with Question 3.d. above.
Response 3G:
Critical regions of the pool structure are identifed on Figure 1.
The design stresses occuring at these locations are given in Table 1.
Table 1 Summary of Stresses Rebar Stress (ksi)
Max. Concrete Critical Horz. Bar Vert. Bar Compressive Stress (ksi)
Location Inside Outside Inside Outside Horz.
Vert.
- 1. West Wall
-0.57 -22.51 46.54
-2.47 1.97 0.55
- 2. North Wall -14.51 1.61
-5.03 30.61 1.42 0.90
- 3. East Wall
-19.77
-3.87
-5.01 31.25 1.78 0.89
- 4. East Wall
-30.11
-7.11
-11.08
-1.66 2.38 1.12
- 5. Basemat
-23.66 0.78
-18.35 14.60 2.09 1.74 Notes:
1.
For the basemat horz. bar is the east-west reinforcing and vert. bar is the north-south reinforcing.
2.
Concrete shear stess is not critical.
The above stresses are well within the allowabic values per i
the acceptance critiera given in Response 3.D.
)
l
H.
Provide description of any changes to be made to the leak monitoring system for the fuel pool.
Response 3H:
No changes are required for the leak monitoring system.
Item 4:
It is not quite clear from the report how the support legs of the racks are_ proposed to be constructed.
provide the following information with respect to the design, construction, and installation of the support legs:
a.
Sketch (or sketches) showing the upper and lower parts of the support feet including where the material transition from SA-217-CA15 to SA-351-CF3 occurs, b.
How is the transition accounted for in the analysis of fuel racks?
Adjusting the support legs during installation would require a c.
long arm wrench.
provide information regarding the installation of racks when an adequate clearance to adjust the supports may not be available.
How is the unevenness (if any) in the installation of the racks accounted for in the analysis?
Response
a,b. Fig. 3.6 (p. 3-13 of licensing report) shows details of the support leg.
The upper part of the support leg is constructed of SA-351-CF3 material and has eight additional gussets.
The upper piece is the female part of the support leg.
The male piece of the support leg, a threaded connection, is made of 3A-217-CA15 material.
In the dynamic analysis, different material allowable stress values are used when checking for structural integrity.
The adjustment tool is shown in the attached JO-drawing.
The c.
use of the tool is independent of clearances.
Therefore, all levelling of the rack can be carried out at installation.
The analysis, therefore, presumes a level rack.
The installation procedure Jp-2481-22 Rev. O was sent under separate cover.
i l
I 4
i
,,Itcm 5:
Tha following information is n:cd:d for us to parform indspendant review of the analysis:
a.
A set of fuel rack and fuel assembly drawings.
b.
All analytical modeling parameters including:
- dimensions
- masses
- spring constants
- gap element properties
- fluid coupling coefficients
- coefficient for restitution for impacts
- multi-racks to rack and multi-racks to pool wall impacts c.
All supporting calculations for model parameters d.
Digitized time histories (OBE & SSE) in three directions, as applicable.
Calculations supporting the assumption that the fuel racks are e.
rigid.
f.
Detailed seismic analysis and results.
g.
Detailed seismic stress analysis and results.
h.
Thermal stress evaluations.
i.
Various limiting rack and multi-rack displacements due to seismic loads.
Response
a.
The following Joseph Oat drawings were sent to Brookhaven National Laboratory under separate cover:
C-8155 Rev. 1 D-8152 Rev. 1 C-8120 Rev. 4 D-8153 Rev. 1 D-8121 Rev. 5 D-8154 Rev. 2 D-8122 Rev. 3 D-8243 Rev. 1 D-8123 Rev. 2 E-8119 Rev. 5 D-8150 Rev. O E-8149 Rev. 1 D-8151 Rev. 1 D-8261 Rev. O b.
Referencing Figures 6.2.1, 6.2.2, 6.2.3, and 6.3.1 of the licensing document, along with Table 6.3.1 (p. 6-28), we provide an attachment which summarizes the spring constant values used for the model.
This attachment is labelled as Table 3.la.
There are four levels of rattling fuel assembly.
The uppermost mass is 12.5% of the total assembly mass; the remaining three rattling masses are each 25% of the total fuel assembly mass, and 12.5% of the total fuel assembly mass is assumed to be attached to the fuel rack base plate.
Fluid coupling coefficients are calculated internally in the dynamic analysis code in accordance with the Fritz model.
[R.J. Fritz, "The Effects of Liquids on the Dynamic Motions of Immersed Solids, " Journal of Engineering for Industry, ASME, 2/72, pp.
167-172.]
Item 5:
(Cont'd)
Response
b.
The analysis considers only a single rack subjected to a 3-D seismic event.
The analysis assumes that all adjacent racks will have equal and opposite local velocities to the rack being studied.
This leads to maximum impact force predictions, c,d Model parameters are calculated internal to the computer code based on inputs of cell configuration and geometry.
A portion of a computer output is provided showing the mass matrix for a full rack with 168 modules computer output for th SSE EW, NS, VERT are also provided.
The data is in "g" units with.01 seconds between each data point in a row, The fuel racks are considered as rigid in the time history e.
analysis by virtue of the fact that the lowest frequencies of vibration of the rack, treated as a cantilever beam, are well above the dominant earthquake frequency.
The seismic forcing frequency is about 5 HZ maximum.
Using an 8 x 14' module bending in the weak direction, we find that the lowest cantilever frequency is 80 redians per seconds.
The inertia of the rack is about the weak axis, and the mass assumed to vibrate (for the purposes of the preceding calculation) is the rack metal mass, all of the mass of water inside the rack plus all of the fuel assembly mass.
f,g Tables 6.1,.6.2 attached summarize maximum loads, displacements, and stress factors (see licensing document p.
6-19) for all of the runs.
These are abstracted from the computer outputs.
All calculations leading to these values are done internal to the code and have been previously verified by hand computations.
h.
The only thermal load of consequence to the rack analysis is to i
examine the effect of an isolated hot cell on the welds.
This is done by considering the effects of heating up a long strip (the cell wall) and retraining the expansion by the edge weld.
We find stresses from this condition to be below the allowable.
i.
See Tables 6.1-6.2 for items f,g.
Table 3.1a (Refer to Figure 3.1)
RACK 8 x 14 12 x 14 5
x 0
.576 x 10 5 Kg(#/in) 7 K,(#/in)
.1 x 10
.1 x 10 7 10 K (#/in)
.113 x 10
.113 x 10 18 f
7 7
K (#/in)
.246 x 10
.244 x 10 g
e a
K,
(
")
.314 x 10
.314 x 10 rad h (in.)
7.25 7.25 H (in.)
169 169 1
1 Hominal gap values for impact springs between fuel assembly l
and cell wall are.151" on each side.
2 Cap values used are the approximate distances between the wall and the girdle bar, or 50% of the spacing between girdle bars and adjacent racks.
For a rack completely surrounded by other racks, the gap for K, springs is
.125".
t
~.
Rosults cro givon horo for a 12x14 ccdulo (tho largost acdule),
and j
~
for a 8x14 configuration (which is a module with the largest aspect ratio) next to the cask pit.
A complete synopsis of the analysis of the 12x14 module subject to the SSE earthquake motions is presented in a summary table labelled as Table 6.1.
Table 6.1 gives the maximum values of stress factors R1 (i 1,2,3,4,5,6).
The values given in the tables are the
=
maximum values in time and space (all sections of the rack).
Table 6.2 gives typical results for the 8x14 rack. The stress factors are defined ast R 1 = Ratio of direct tensile or compressive stress on a net section to its allowable value (note support feet only support compression)
R 2 = Ratio of gross shear on a not section to its allowable value R 3 = Ratio of maximum bending stress due, to bending about I
the x-axis to its allowable value for the section Rg = Ratio of maximum bending stress due to bending about the y-axis to its allowable value R 5 = Combined flexure and compressive factor (as defined in 5.3.1e above)
=
R 6 = Combined flexure and tension (or compression) factor (as defined in 5.3.1f above)
As stated before. the ellevable value of R _(1 =1.2.3.4.5.4) 1 is 1 for the OBE condities and 2 for the SSE earthquake.
The results given in the tables are for the SSE earthquake and have been given in the licensing document. The maximum stress factors (Ri) are below the limiting value for the SSE condition for all sections. It is noted that the critical load factors reported for i
the support feet are all for the upper segment of the foot and are i
l to be compared with the limiting value of 2.0.
i l
Tablos 6.1 - 6.2 also prosent results chich are usod to shoo that significant margins of safety exist against local deformation of the fuel storage cell due to rattling impact of fuel assemblies and against local overstress of impact bars due to inter-rack impact.
The tabular results shown assume that the rack metal thicknesses are based on a total metal thickness of 0.06 inch in the cell structure area.
l l
l
Table 6.1 BYROM RACKS -
SUMMARY
OF RESULTS 12 x 14 (Surrounded by other Racks) l MAX. DISP. D, NAX. DISP. D MAX. FLOOR LOAD MAX. FLOOR y
RUN #
REMARKS (In.)
(In.)
(4 feet) (f) 1 Foot (f) V/H" C012 y=
.8
.1683 (IMP)
.0877 4.538 x 10 1.996x10 /1.304x10 5
5 5
SSE, full C016 p=
.8
.1591 (IMP)
.0528 2.348 x 10 1.159x10 /69374 5
5 l
SSE, half full, pos. x C013 p=
.2
.1534 (IMP)
.0868 4.915 x 10 1.499 x 10 /29525 5
5 SSE full C017 p=
.2
.1172
.0724 2.386 x 10 1.018 x 10 /18065 5
5 SSE half full, pos. x 1
l C014 p=
8
.1425 (IMP)
.0452 6.631 x 10" 4.129 x 10"/30467 10% full C018 p=
.8 C1
.1722 (IMP)
.0843 4.633 x 10 1.985 x 10 /152457 5
5 edge rack full load l
C015 y=
.2
.1291 (IMP)
.1191 87830 4.050 x 10"/6835 10% full C019 p=
.2
.1317 (IMP)
.1470 4.289 x 10 1.862 x 10 /31884 5
5 C1 edge rack (full)
V = Vertical Direction H = Horizontal Direction I
L
1 Table 6.1 (continued)
. BYRON RACKS LOAD FACTORS (UPPER VALUES FOR RACK BASE -
FUEL ASSEMBLY LOWER VALUES FOR SUPPORT FEET)
TO CELL INTER RACK l
RUN #
IMPACT LOAD IMPACT LOAD R t R 2 R 3 Rg R 3 R 6 I
C012 55190 43670
.083
.108
.176
.323
.356
.405 l
.545
.594
.662
.853 1.032 1.163 l
l C016 36270 34970
.044
.052
.068
.187
.213
.243
.316
.322
.269
.410
.570
.619 i
.092
._040
.099
.224
.267
.306 i
C013 53970 27980 q
.416
.136
.202
.179
.536
.558 i
C017 22800
.045
.015
.062
.157
.176
.199
.281
.084
.106
.120
.302
.306
.011
.019
.060
.049
.071
.082 C014 17570 17530
.115
.136
.199
.142
.239
.270 i
.088
.123
.142
.285
.351
.400 C018 55160 46670
.529
.702
.429 1.011 1.189 1.342 t
C015 19930 4310
.015
.006
.058
.049
.064
.074 3
.095
.031
.051
.055
.128
.134 J
.081
.033
.088
.216
.258
.295 C019 53480 5825
]
.503
.137
.168
.232
.570
.595
Table 6.2 BYROM RACKS -
SUMMARY
OF RESULTS E2 RACK 8 r. 14 MAX. DISP. D, MAX. DISP. D MAX. FLOOR LOAD MAX. FLOOR
- y RUN #
REMARKS (in.)
(In.)
(4 feet) (#)
1 Foot (#)
Vertical / Shear C020 p=
.8 SSE
.1731 (IMP)
.0691 3.149 x 10 1.495 x 10 /98932 5
5 full load CO23 p=
.8 SSE
.6263 (IMP)
.1324 7.055 x 10 "
5.415 x 10"/38994 10% filled no cent. offset C021 p=
.2
.4567 (IMP)
.0877 2.765 x 10 1.321 x 10 /25827 5
5 SSE full load C024 p=
.8 SSE
.4767 (IMP)
.1127 2.345 x 10 1.286 x 10 /66013 5
S 50% neg. x-axis CO25 p=
.2, SSE
.1407 (IMP)
.1043 1.532 x 10 8.485 x 10"/16513 5
50% neg. x-axis CO22 9=
.2
.8012
.1312 1.129 x 10 7.863 x 10 /13382 5
4 SSE 10% f u l.1 VERT Vertical reaction
=
SHEAR = Horizontal Reaction
Table 6.2 (continued)
BYRON RACKS E2 RACK 8 x 14 LOAD FACTORS (UPPER VALUES FOR RACK BASE -
FUEL ASSEMBLY LOWER VALUES FOR SUPPORT FEET)
)
TO CELL INTER RACK RUN #
IMPACT LOAD IMPACT LOAD Rg R 2 R 3 Rg R 3 R 6 4
.088
.115
.233
.278
.366
.419 10 "
4.812 x 10 C020 2.974 x
.412
.392
.674
.555
.878
.996 l
l CO23 1.265 x 4
.018
.024
.063
.096
.141
.163 10 "
1.398 x 10
.135
.160
.178
.286
.358
.407
.0M
.032
.133
.283
.343
.391 CO21 3.767 x 10 "
6039.
.359
.119
.179
.464
.637
.704 4
.065
.079
.147
.221
.316
.362 10 "
4.386 x 10 CO24 2.521 x
.350
.309
.422
.482
.602
.678 CO25 2.720 x 10 "
1.485 x 10 "
.042
.019
.097
.182
.217
.249
.229
.073
.111
.275
.382
.422 4
.021
.011
.107
.098
.171
.197 10 "
2.806 x 10 CO22 1.609 x
.186
.046
.093
.258
.285
.323 i
=
' Item 6:
Specific questio'ns related to the analysis:
Provide' justification for ignoring the flexural rigidity of a.
I
. fuel-assemblies and modeling them as-five separated " rattling" Provide information on the flexural rigidity of fuel masses.
assemblies.
Also discuss'the manner in which the flexibilities of the multi-element fuel assemblies and cells are accounted for in the rack module analysis.
4 b.
How is the vertical mass of the fuel assemblies accounted for in the model?
How is the mass of the water within the fuel accounted for in c.
-the model?
d.
Have the effects of hot gaps and cold gaps been considered?
e.
The model in Figure 6.7 shows the gap elements between the rack module edges and fixed boundaries are used to simulate inter-rack impacts.
This does not simulate the potential increase of gaps due to sliding of a row of rack modules and development of higher impact velocities with larger gaps.
Provide further justification that such worst case effects have been accounted for, f.
Demonstrate that inter-rack impacts at mid-height elevations are not possible.
g.
Explain the significant frequency difference between the EW and the NS SSE time histories shown in Figures 6.1 and 6.2.
Response
a.
The rack inertia property for banding is calculated to be at least 65000 in4 Since all cells are connected together to provide a complete grid structure, comparisons of individual fuel cell inertia should be made with the above value.
Assuming a fuel assembly as a solid bar 9" x 9" in cross section gives an inertia, per assembly, of 547 in4 This leads us to neglect the assembly stiffness in the calculation and consider only the mass distribution.
The rack inertia is calculated by considering an equivalent rectangular gridwork with thicknesses equal to the cell wall, and having the same metal area as the rack structure cross section.
This easily leads to a determination of the necessary inertia characteristics for mass calculations.
b.
All of the fuel assembly mass is assumed to move with the rack base in the calculation of vertical inertia contributions.
c.
Mass of water within the racks is automatically included when the Fritz coupling model is used.
The fluid mass effects are felt both in an "added mass" type term; and in a " hydrodynamic mass" type term which is inversely proportional to the annular gap.
l ve-
,,,-e.--.,,,-,..,
-,,n,,n
._,.-n_.,
,---..-...m
,.._...,._,,-..-n_,.
Item 6: (Cont'd)
Response
i d.
All calculations have been based on nominal gaps.
On some previous jobs, we have had occasion to vary the assembly - cell wall gap to a small degree; results showed that final values were not sensitive to small variations in gap dimension.
Certain cases are run with larger gaps on one or more wall e.
(simulating a rack at the edge of the group).
This generally turns out to be the limiting case since the rack-rack gap also affects the hydrodynamic mass that is present due to having the rack itself considered moving in an enclosed area (surrounded by other racks).
Since the use of a hypothetical fixed boundary (based on symmetry considerations) is based on the conservative assumption of out of phase motion of all adjacent structures, we feel results are certainly conservative and are bounded by the extreme case of a vibrating edge rack, f.
Mid height rack to rack impact is not considered since the rack is essentially rigid.
The rack will never deform enough to impact at mid height.
The girdle bars at the top prevent rack cell walls from ever coming in contact except at the top and at the base plate.
The exciting frequencies are not high enough co cause other than rigid body motions of the rack, g.
Frequency difference of about 2 HZ is due to our strenuous efforts to insure statistically independent horizontal movements from time history data developed from the specified response spectra.
Different envelopes were used as input data to the time history generation code.
l 1
f
Item 7:
Provide the parameters and constants used in the analysis for impact loading due to the drop of a fuel assembly.
Also, provide a summary of ductility ratios utilized to absorb kinetic energy in the tensile, flexural, compressive and shearing modes.
Provide typical calculations indicating the input constants, equations used, and the results of the impact analysis.
Response
Drop analysis used sample fluid model to estimate drag acting on cell assembly, Numerical analysis is then used to calculate final velocity of assembly at the bottom of the rack.
This can be used to l
get the input kinetic energy which must be absorbed by the base plate.
We show that shear stresses that develop are not sufficient to cause a " push out" of the base plate.
A draft of typical calculation is attached.
No ductility factors are used.
l I
Let:
weight of fuel assembly (1616 lb)
W
=
h height above base plate (201 in)
=
side of (square) effective solid cross-section of a
=
fuel assembly = 8.548" thickness of base plate (0.625 in)
T
=
Yield stress of base plate material (23,150 psi)
Y
=
6 distance that fuel assembly penetrates into the
=
base plate Figure 7-2 shows the final deformed configuration of the base plate with fuel assembly sitting on it.
The work dissipated in plastic deformation is F6 where F is the average resisting force exer'id on the fuel assembly by the base plate.
Using the model which has proved accurate for the punching of slugs out of plates- (see Paul and Zaid (1958)] the force F may be conservatively estimated by F=
(
Y) 4aT
where /3 Y is the yield stress in shear according to the distortion energy theory of yielding.
l Then the work dissipated in plastic deformation is
[3 F6=
YaT6 Upon equating this work to the kinetic energy gained during the fall, we find i
YaT6 = b y 2
/3 2g so that 3
W f
6=
4 YaT 2g To calculate the final velocity of a body dropping through a channel, we account for virtual mass, gravity and fluid drag.
We assume that the virtual mass is equal to the buoyant
- mass, and that the drag coefficient is based on exposed frontal area of the fuel rods.
The governing equation for a mass element in free fall subject only to gravity and drag ef fects is C*
2,.(g _ g )g (M + My) v+
p,A v
2 where Cg = effective drag coefficient due to all con-tributing effects, and My virtual mass of object.
=
If c = Wy/W, then I
D^*
9 g = ( 1-c) gw
( 1+ c) 2
( 1+ c)
9 subject to v = 0 at x = 0 g at x = h v=v In finite difference form Av t,y = At [ 1-c g
-w D A*'i C
]
1+ c 2w
( 1+ c )
t v,y = vi+
Av, y g + (v +v,y)At/2
- x,y = x g
t t
g g
For a rod like body with characteristics similar to a fuel assembly, Cg a 1.0 + A where A represents the incremental increase in ef fective drag coef ficient due to the fluid being confined in a
narrow channel.
Consider Figure 8-3.
A*v = A v 3; A 3
cell f*A'3 3
Therefore v3= uv;vg= ulv where A
A U=
- ut=
A 3 A,yy c
Assuming that the expansion at A 3 is to a very large area, then the pressure p 3 is essentially equal to the fluid pressure outside of the cell.
Neglecting any depth effects, an energy balance yields the pressure difference across the fuel assembly as 2
o ov ap = -*
(v3 2_y 2 W
2 (9
u1 2)
)
g 2
2 i
J
as long as u 2 ut.
The effective incremental resist-ing force due to cell geometry is AF = apA* so that 2
^
A=
U ut
[l-A 3 /A
=
cell A 3 The BASIC program on the next page has been written for an IBM micro computer.
The accuracy of the results can easily be verified by considering the drag free case with vg = /2gh for any typical set of input data.
For a fuel cell dtop to the base of the rack through a distance of 201",
the final velocity at impact with the rack base is V1 = 202 in/sec Therefore, the maximum depth of penetration is 2
H V I
.443 x 1616 x 202 6=
3
=
YaT 2g 23150 x 8.548 x.625 x 386.4 x 2 6=
.306" <
.62'"
Therefore, the base plate will not be penetrated by the drop.
The center-to-center distance between adjacent storage cells is not dependent on the presence or absence of support from i
the base plate.
The purpose of the above calculation is only to show that there is no danger to the liner.
In the event of a dropped fuel assembly, it is correct to say that the base plate will separate from the tube in the immediate vicinity of the affected tube.
While this would result in base plate plastic bending, it would not affect center-to-center spacing since there would be ao effect on the welds between adjacent tubes nor on the base plate-to-tube welds away from the immediately vicinity of the dropped assembly.
The base plate, even with plastic bending occurring, will not touch the liner floor in the event of a dropped fuel assembly hitting the base plate.
,e--
e-
--,-,,y w
Drop to Top of Rack acell =
78.32291 astar=70.9638 a3 50.26544
=
cell length =165 weight =
1616
. time step =.001 drop distance = 36 eps=
.268357 cd=
1 x1=
36.0581 v1=125.3354 Drop With No Virtual Mass Effect and Cd= 0 acell=
78.1456 astar=70.82908 a3 28.27431
=
cell length =165 weight =
1616 time step =.001 drop distance = 36 eps=
0 cd=
0 1=
36.05587 v1=166.9243 Drop to Base of Rack acell=
78.32251 astar=73.06831 a3 28.27431
=
cell length =165 weight =
1616 time step =.001 drop distance = 201 eps=
.2763172 cd=
6.8081 x1=
201.1655 v1=202.3124
l 10 REM fuet r'a c < c r'c o c a l c u l m i c^'s INPUT cel1 dim.,gao,nule **acrus:CD.GC,h9
-AC=CD*CD:AS=(CD-GP)*(CD-GCir a 3 = 2.1 4159*"7+W4 13 tc4 INT" acell=",AC,"astar=".A5 14 _DRIN7
'a3= ",A3 15 LM=(AS/A3)"2*(1-(A3/AC)^2) 16 PRINT "Im =
._M 17 PRINT."tnout acove v a l..t e f o r' 1m c r' teout 0 if d e'a q co=f =*.r" 18 INPUT LM CD=1'+e_M 20 INDUT "1,w,dt,xf";L,W,DT,X:
21 LPRINT " fuel length =";L 22 LPRINT " weight =",W," time steo=",DT 23 LPRINT " drop distance =",XC 24 INPUT" READ IN FUEL SOLIDITV";F9 25 EP=FR*AS*L*64/1728*1!/W 23 LPRINT "eps=",EP LPRINT "cd=",CD 40 C2=(1-EP)/(1+EP) 50 C1=64'/1728'*CD/2!*AS/(W*(1+EP))
60 X1=0 70 Vi=O X=0 CO V=V1 X=X1 90 DV=DT*(386.4*C2-Cl*V*V) 100 V1=V+DV:X1=X+.5*(V+V1)*DT 105 PRINT "x1=",X1,"
v1=",V1 109 ARINT "x1=",X1,"
v1=",V1 110 IF X1)=XF THEN GOTO 140 E'.SE GOTO 80 t PRINT "x1=",X1,"
v1=",V1
- > _C9 TNT "x1=",X1,"
v1=",V1
. O LORINr" ":LpAINT" 200 END
-a-
/
n
,/
a.
F W
FUEL' ASSEMBLY 6-U YK
.,/
a
'p)r
(...
i/.
ei r,
j BASE.
. P L A S TIC PLATE DEFORMATION i
FIGURE 7.2 i
l j
i i
l i
l e
3
'A y
i le c
0 F
A g
v
=
S r
3 7
E R
/
p U
G a
I g
F A
}
e
/
'e w
4 A.
e i
4
!,i ii1
..I, 1l l21 4
ll!
4Ii 4iliiI 4
4
)
Item 8:
l Provide the considerations given regarding the potential impact on the functionality of fuel rack modules due to bowing and localized deformations of fuel assemblies and fuel rack cells.
Response
Sufficient fuel assembly - cell wall clearance is provided so that bowing and localized deformation is not a problem.
We show that the cell wall has sufficient strength to withstand any seismic impact load without causing any safety related affect.
The dynamic analyses gives predictions of maximum assembly to cell wall impact loads.
Thereafter, our concern is simply that the cell wall does not undergo significant plastic deformation.
SB/dg/0290B
Request:
9.
Provide the estimated occupational dose expected from performing the " dry" reracking of the Byron spent fuel pool.
These doses should be quantified (e.g., 10 mrem, 0.1 mrem, less than 0.1 mrem) and include the following:
a.
occupational dose for each phase of the SFP modification; b.
the basis for the estimate (methodology), including dose rates and manpower; c.
maximum individual dose expected.
10.
Provide the information in 9, above for the " wet" reracking, additionally addressing doses to divers.
(NOTE:
If wet reracking is not addressed, the safety evaluation will consider dry reracking only.)
Response
Byron's spent fuel pool is presently dry. All but eight or nine of the old NUS supplied spent fuel storage racks (i.e., those which provide low density fuel assembly storage) have been removed from the pool. There are several scenarios for removal of the remaining racks and installation of the new racks which allow high density fuel storage.
These scenarios are dependent upon the swiftness with which new racks are procured and licensed.
Conceivably, the remainder of the old racks can be removed and the new racks installed prior to water being added to the spent fuel pool (i.e., " dry" reracking).
In this scenario there are no radioactive sources within the fuel pool and thus the occupational dose due to reracking will be minimal.
Another likely scenario is that the remainder of the old racks will be removed and the new racks installed after water has been added to the fuel pool and spent fuel has been stored therein (i.e., " wet" reracking).
For the purposes of estimating occupational doses during reracking, two scenarios will be considered.
The first is the dry reracking scenario and the remaining scenario is for wet reracking.
The wet reracking scenario assumes all work (i.e., removal and installation) takes place after spent fuel is introduced into the spent fuel pool.
The following assumptions are used in determining the occupational doses involved with the reracking operation.
- The dose rate environment to the work area outside the spent fuel pool is assumed to be less than or equal to two mrem /hr (Figure 12.3-32 of the Byron /Braidwood Station FSAR).
- The dose rate to a diver submerged within the spent fuel pool water when spent fuel is located within the pool is assumed to be five mrem /hr.
This value is simply two times the maximum expected dose rate (i.e., 2.5 mrem /hr at the spent fuel pool water su 9 ace (pg. 9.1-26 of the Byron /
Braidwood Station FSAR, Amendment 43, September 1983).
r Response (Cont'd):
- It is assumed that the diver is at a sufficient underwater distance from the stored spent fuel assemblies such that the direct dose rate to the diver front the spent fuel assemblies is insignificant compared to the immersion dose rate of five mrem /hr due to radioactive sources suspended in the spent fuel pool water.
- When there is no water within the spent fuel pool (i.e., " dry" reracking) or when there is water but no spent fuel located within the pool, the dose rate within the pool will be assumed to be equal to the dose rate environ-ment outside the pool under normal operating conditions i.e., two mrem /hr.
This is a conservative assumption since dissolved activation products due to stored spent fuel is a major contributor to the dose rate environment outside the pool.
- Twenty-three new spent fuel racks are to be installed (pg. 2-1, " Licensing Report on High Density Spent Fuel Racks for Byron Units 1 and 2," NRC Docket No. 50-454, 50-455, July 25,1986).
- The total construction time for installation of the new racks is about 24 man-hours (i.e., 3 man-days) per rack (in the refueling building - top floor).
Of this, about eight man-hours (one man-day) per rack involves use of a diver within the pool installing the new racks.
The three man-day per rack figure also includes the crane operator who will be maneuvering the racks.
(R. Salsbury calling A. G. Klazura, " Construction Time Involved with the Spent Fuel Pool Modification." Memorandum of Telephone Conversation dated 11-19-86).
Occupational Dose Analysis 1.
Dry Reracking The occupational dose associated with dry reracking is about 1.5 man rems and is determined as follows:
Dose due to removal of nine existing (i.e., old) storage racks:
(9 racks)x(3 man-days / rack)x(2 mrem /hr)x(8 hr/ man-day) 432 mrem
=
Dose due to installation of 23 new storage racks:
(23 racks)x(3 man-days / rack)x(2 mrem /hr)x(8 hr/ man-day) = 1104 mrem Total dose for dry reracking:
432
+ 1104 1536 mrem or
~ 1.5 man Rems During dry reracking the phases of work were not treated separately because a constant dose rate value of two mrem /hr was assumed for all phases of activity.
The assumed dose rate value of two mrem /hr is conservative because the dose rate to the area immediately around the spent fuel pool would be expected to have a dose rate considerably less than two mrem /hr when there are no spent fuel assemblies within the spent fuel pool.
The actual dose rate will be only slightly above that due to natural background.
The
Occupational Dose Analysis'(Cont'd) calculated dose value of 1.5 man rems is thus a conservatively determined upper limit to the exposure associated with dry reracking.
If it is-assumed that a particular individual is always present during the reracking operation and that the reracking operation requires the presence of three people, then the maximum individual dose expected would-be less than 0.5 rems.
2.
Wet Reracking Scenario:
This scenario assumes that nine old racks are removed and 23 new racks are installed after spent fuel has been placed in the spent fuel pool.
The occupational dose associated with the wet racking scenario is about 2.3 man-rems.
This dose is determined as follows:
Dose to diver during removal of nine old racks:
(9 racks)x(1 ay)x(5*[*r)x(8
)=
360 mrem ma d
Dose to non-divers during handling of nine old racks:
(9 racks)x(2 S)x(2*[*)x(8 da ) = 288 mrem ma" ra k r
Dose to diver during installation of 23 new racks:
ma d (23 racks)x(1
)x(5 "I*r )x(8 d )=
920 mrem k
Dose to non-divers during handling of 23 new racks:
(23 racks)x(2 ays)x(2 *hr )x(8 da ) = 736 mrem man Total dose for second wet reracking scenario:
360 288 920
+ 736 2304 mrem or ~ 2.3 man Rems If it is assumed that one diver handles both the removal of the old racks and installation of the new racks then this would result in the maximum individual dose which would be 360 + 920 = 1280 mrems.
Request:
11 Identify, for " dry" reracking (and " wet" reracking if planned):
1.
radwaste sources (e.g., racks, filters) volumes, and types (e.g.,
solid, liquid gaseous) expected from:
Request (Cont'd):
a.
the reracking operation, and b.
from subsequent spent fuel pool operations; and 2.
the expected occupational dose increases resulting frora increased spent fuel storage (e.g., compare pre-mod radwaste doses and post-mod radwaste doses).
Response
ll.la Radwaste Sources Resulting from the Reracking Operation.
Dry Reracking There will be no expected radwaste sources associated with dry reracking.
During dry reracking the spent fuel pool contains no spent fuel assemblies and is empty of water. Thus, there are no radioactive sources available to produce contamination.
Wet Reracking The radwaste sources associated with wet reracking (i.e., those above and beyond the specific activity normally expected in the spent fuel pool water and air above the water) would primarily be of the solid and liquid types.
The solid radwaste sources would include contaminated wet suits, diving gear, and tools which couldn't be sufficiently decontaminated.
It is assumed tools, diving gear, etc., will be reused as much as possible to reduce the quantity of solid radwaste.
The primary source of liquid radwaste would be water used to decontaminate.
This liquid radwaste is expected to have extremely low specific activity and would most likely be passed through the floor drain system.
The volume of liquid radwaste produced depends on the quantity of items contaminated and the amount of water required to sufficiently decontami-nate.
It is impossible to specify volumes of liquid radwaste expected.
For the situation in which spent fuel is stored in the spent fuel pool prior to the removal of all the old spent fuel racks, the old spent fuel racks upon removal from the pool would have to be sprayed down to remove contamination.
This fluid (most likely water) which comprises the spray would be treated as liquid radwaste.
It would have a very low specific activity and would most likely be directed to the floor drain system. Any racks which could not be sufficiently decontaminated would have to be treated as solid radwaste.
A potential source of gaseous radwaste during reracking is noble gases which could be released from solution due to agitation of the water by the divers.
Most of the noble gases which are able to be released from the fuel will be released upon depressurization of the reactor vessel or during transit to the spent fuel storage pool.
The gases in solution alluded to here are daughter products of halogens. Any gases released from solution due to agitation of the water would be handled by the fuel pool gaseous exhaust system. Agitation of the water doesn't affect the quantity of radioactive gases from the spent fuel pool which have to be handled by the radwaste system, it simply quickens their release.
Response (Cont'd):
ll.lb The radwaste sources resulting from subsequent spent fuel pool operation for the situation involving the new spent fuel racks (i.e., in the high density storage configuration) are expected to be virtually identical to the radwaste sources which were expected from the fuel pool operation for the situation involving the old spent fuel racks.
Radioactivity in spent fuel pool water is due to mixing of some reactor coolant water with the spent fuel pool water during refueling operations, to the release of surface crud on the spent fuel assemblies, and to fission product leakage from defective spent fuel elements (pg. 4-11 of NUREG-0575, Vol. 1, Executive Summary Text, Project No. M-4, August,1979).
Modification to the storage capacity of the spent fuel pool has no impact on the fuel transfer operation.
The amount of reactor coolant water mixed in with the spent fuel pool water during fuel transfer operations is unaffected by the modification as is the reactor coolant water contribution to the radioactivity in the spent fuel pool water.
Modification to the storage capacity also has no affect on the amount of crud on the surfaces of spent fuel assemblies nor on the crud contribu-tion to the radioactivity in the spent fuel pool water.
The spent fuel pool storage rack modification for Byron increases the number of spent fuel assemblies which can be stored within the pool from_1060 assemblies to 2940 assemblies.
The greater number of fuel assemblies means there is greater potential for leakage from defective fuel elements.
This could lead to increased radioactivity in the spent fuel pool water.
Studies have shown however, that the leakage of radioactivty into spent fuel pool water from defective fuel elements which have been stored for several months is nearly nonexistant (pg. 4-12 of NURGE-0575, Vol.1, Executive Summary Text, Project No. M-4, August,1979).
For the Byron spent fuel pool modification, the spent fuel storage capacity is increased from nine years to about 24 years.
The additional fuel elements stored in the additional space provided by the modified spent fuel pool storage rack configuration are thus effectively nine years old. The effect on sources due to the increased capacity of the spent fuel pool is insignificant.
11.2 Cccupational doses due to transfer and storage of spent fuel are the doses received during transfer of spent fuel to the storage pool and arrangement of the fuel within the storage pool, doses above the pool water surface due to radioactive sources within the spent fuel pool water, doses due to radwaste sources generated during the transfer and storage of spent fuel, doses above the water surface due to radiation shine from the stored fuel assemblies, and doses due to radiation shine through the spent fuel pool walls.
As stated in response toll.lb, modification of the spent fuel pool storage capacity has no impact on the fuel transfer operation and no significant impact on radioactivity in spent fuel pool water nor on the quantities of radwaste generated.
The spent fuel pool storage modification scheme thus has no affect on the occupational doses due to transfer of spent fuel, no affect on the doses above the pool water surface due to radioactive sources within the spent fuel pool water, or no affect on the doses due to radwaste sources generated during the transfer and storage of spent fuel.
e Response (Cont'd):
The radiation doses above the pool water surface due to radiation shine from the stored fuel assemblies are an immeasurably small fraction of the doses due to natural background radiation.
The higher density storage of spent fuel assemblies allowed by the modification will result in doses above the pool water surface (due to radiation shine from the stored fuel assemblies) which are still only an immeasurably small fraction of the doses due to natural background radiation. The spent fuel pool storage modification scheme has no affect on the occupa-tional doses above the pool water surface due to radiation shine from the stored fuel assemblies.
Table 7.3 (of the " Licensing Report on High Density Spent Fuel Racks for Byron Units 1 and 2," NRC Docket No. 50-454, 50-455, CECO, July 25, 1986) lists the dose rates expected to the areas around the spent fuel pool due to radiation shine through the fuel pool walls and the high density storage configuration. These are the maximum expected dose rates due to freshly discharged fuel which is located as close as the storage configuration allows to the pool walls.
The dose rate values in Table 7.3 of the stated report will be reduced by a factor of six, 60 days after the fuel is placed within the pool.
The east wall of the spent fuel pool is 5.5 ft. thick concrete and separates the spent fuel pool and the fuel transfer canal.
The north and south walls of the spent fuel pool are 5 ft. thick concrete. They lie between the spent fuel pool and pipe penetration areas.
The west spent fuel pool wall is 6 ft. thick.
It lies between the spent fuel pool and the fuel pool heat exchanger area.
The pipe penetration areas and the fuel pool heat exchanger area have normal operation radiation zones of 20 to 100 mrem /hr.
These areas are thus radiation areas.
Access to them is controlled. As shown in Table 7.3 of the stated report, the dose rate (for the modified storage configuration) adjacent to the north spent fuel pool wall is 54 mrem /hr and the dose rate adjacent to the west wall is 4 mrem /hr.
The dose rate adjacent to the north spent fuel pool wall due to the old storage configuration was calculated to be about 23 mrem /hr and the dose rate adjacent to the west wall was calculated to be about 0.24 mrem /hr.
The differences in dose rates when comparing the old to new high density storage configura-tion are mainly attributed to the reduced distances between the fuel assemblies and the spent fuel pool walls which are necessary to accommodate the increased storage capacity.
As previously stated the areas outside the north, south, and west walls of the spent fuel pool are not normally occupied and access to the areas is controlled. Additionally, the dose rates outside these walls reduce drastically with time.
In light of these considerations, the high density configuration for storing spent fuel would have no significant impact on the occupational dose due to radiation shine through the spent fuel pool walls.
Reguest:
12.
Quantify the gaseous effluent releases (i.e., Kr-85) and compare the releases expected due to increased spent fuel storage quantitatively with pre-mod and post-mod annual spent fuel pool and overall plant releases- (actual or calculated).
Response
The storage capacity of Byron's spent fuel pool was originally intended to accommodate storage for nine years worth of spent fuel assemblies resulting from operation of units 1 and 2.
The modified storage capacity is intended to increase storage capacity another 14 years.
The added storage space is effectively used to store fuel assemblies which have been removed from the core for at least nine years.
As stated in Section 7.2 of the " Licensing Report on High Density Spent Fuel Racks for Byron Units 1 and 2" (NRC Docket No. 50-454, 50-455, July 25, 1986),
the aged fuel in the expanded storage capacity will not contain significant amounts of radioactive iodine or short lived gaseous fission products, since these would have decayed during the storage period.
Gases contained within the voids of defective spent fuel elements tend to leak out of the fuel elements quickly during depressurization of the reactor vessel and subsequent handling of the fuel.
The gases which are within the fuel pellet matrix and not in the void have a very low diffusion rate (Section 4.2.2.2 of NUREG-0575, Vol. 1, August,1979).
Thus, the release rate for ga5es from defective spent fuel elements in storage, particularly the noble gas Kr-85 which is the most prominent, is expected to be extremely low in comparison to the most recently stored fuel elements.
The increased spent fuel pool storage capacity will increase the yearly quantity of Kr-85 released from the spent fuel pool into the atmosphere by no more than 3.6 Curies / year per unit.
In obtaining this value it was conservatively assumed that:
- The entire auxiliary building gaseous release rate per year for Kr-85 is due to fuel handling building sources.
(The fuel handling building exhaust system is part of the auxiliary building HVAC system). The expected annual release rate of Kr-85 ventilated from the auxiliary building is 2.0 Curies /
year per unit (Table 11.3-6, Byron /Braidwood FSAR, Amendment 20, April, 1979).
- The value of 2 Curies / year released via the auxiliary building ventilation system is assumed to be representative of the release for the spent fuel pool with the low density storage capacity.
- The spent fuel pool activity released per year is directly proportional to the ratio of the number of fuel assemblies stored in the spent fuel pool.
This assumes that the activity released per assembly is not reduced with time of storage and is thus a very conservative assumption.
(Maximum storage capacities were used in applying the ratio.)
As seen from Table 11.3-6 of the stated FSAR, the total airborne Kr-85 released per year per unit for the entire plant is 700 Curies. The calculated estimate of Kr-85 increase per year (i.e., 3.6 Curies) due to the modified spent fuel pool storage capacity is less than one percent of the total Kr-85 released per year per unit for the entire plant.
Request:
- 13. Provide a quantitative assessment of the impact of releases identified in Question 12, above on individual and population doses offsite.
Response
The response to request numberl2, states that the additional Kr-85 released to the atmosphere as a result of the increased capacity for the spent fuel pool i-is less than one-percent of the Kr-85 released for the entire plant (one unit).
This is arrived at using very conservative assumptions. The predominant gas released from the stored spent fuel is expected to be Kr-85.
It follows that the increased storage capacity modification will result in less than one percent increases to the offsite doses.
5 4
i
.-__.-.,____mm.,,
.-N
'E Request:
14.
Compare the doses from 9 and 10above with the overall doses projected (or actually experienced) for spent fuel operations in the Byron (or a similar plant) SFP and with doses experienced or projected for overall plant operations.
Response
Typical doses experienced at the Zion nuclear power station (a PWR) due to refueling operations (i.e., removing and replacing about 1/3 of the core) are around 4.5. man rems.
The dose of 1.5 man rems acquired during dry reracking as determined in response number one is conservatively estimated as 1/3 the dose expected during a single refueling operation.
The actual dose is expected to be not mm;h greater than that acquired from natural background radiation.
The dose ocquired due to wet reracking was determined in response number two to be about 2.3 man rems.
This is slightly more than half the dose expected during a single refueling operation.
The yearly radiation exposure associated with routine operation of the Zion station lies around 700 to 750 man rems.
The calculated doses associated with dry and wet reracking are less than one percent of the yearly plant exposures.
The reracking operation (even if performed under " wet" conditions when following ALARA practices) results in doses which are on the order of those received during a single refueling operation.
There doses add insignificantly to the expected yearly plant exposures.
((
a Request:
^ 15.
Verify that,no changes to the SfP ventilation and SFP water cleanup systems will be required for radiological purposes (e.g., need for increased capacity, higher flow rates, additional components, revised design or layout).
Response
As stated in response number two, radioactivity in the spent fuel pool water is due to mixing of some reactor coolant water with the spent fuel pool water during refueling operations, to the release of surface crud on the spent fuel assemblies, and to fission product leakage from defective spent fuel rods.
Studies have shown that the leakage of fission products into spent fuel pool water from defective fuel rods which have been stored.for several months is nearly nonexistant. The increased storage capacity enabled by the modified spent fuel pool storage configuration effectively handles fuel which has been out of the core at least nine years.
As such, the radioactivity in spent fuel pool water due to additional storage of fuel assemblies made possible by the modification is insignificant in comparison to the radioactivity contribution due to the most recently stored fuel assemblies.
The existing water cleanup system (i.e., that designed to accommodate the premodification spent fuel pool capacity) is sufficient for the modified. spent fuel pool with incrased storage capacity.
The gas'es released to the atmosphere from the modified spent fuel pool are expected to be insignificantly different in types and quantities from those expected to be released from the premodified spent fuel pool (see response number four).
Even upon application of conservative analysis, the amount of Kr-85 expected to be released per year is less than one percent of the total Kr-85 released for the entire plant (Kr-85 is the principal gas expected to be released from stored spent fuel).
The effect of the spent fuel pool modification on offsite doses due to atmospheric releases is insignificant when compared to the expected releases from the premodified
., spent fuel pool.
The existing spent fuel pool ventilation system is sufficient.
?
]